In the boiling water reactor (BWR), output power can be controlled by changing a core flow and thereby changing a steam ratio (void fraction) inside a boiling reactor core.
However, it is known that depending on the core flow and other operating conditions, neutron flux distribution and liquidity in the reactor core are destabilized by delayed transportation of voids and a negative feedback effect caused by negative void reactivity coefficients in the reactor core.
There is concern that occurrence of such a nuclear thermal hydraulic destabilization phenomenon may result in considerable oscillation of output power and flow rate, which may deteriorate cooling characteristics in terms of fuel rod surface temperature and may damage the soundness of fuel rod cladding tubes.
Accordingly, in designing fuels and reactor cores for the boiling water reactor, the nuclear thermal hydraulic stability is analyzed to produce a design that gives sufficient margin to stability so as to prevent such an oscillation phenomenon from occurring in any of the expected operating ranges.
In such a range where deterioration in nuclear thermal hydraulic stability is expected, limited operation is preset for safety. Nuclear reactors of some types are provided with a safety setting so that in the unlikely event where the nuclear reactor reaches the operation limited range, output power is lowered by insertion of control rods and the like so that the nuclear reactor can get out of the operation limited range.
There are a large number of nuclear power plants which allow, from a viewpoint of Detect and Suppress, power oscillation phenomena while accurately detecting the power oscillation phenomena attributed to nuclear thermal hydraulic destabilization and suppressing the oscillations before the fuel soundness is damaged.
Accordingly, a power oscillation detection algorithm with use of dedicated detection signals for detecting the power oscillation phenomenon, which is referred to as OPRM (Oscillation Power Range Monitor), has been proposed (e.g., Patent Literature 1).
There is also known a technique to analyze principal components in an oscillation based on a plurality of nuclear instrumentation signals, extract independent components different in oscillation modes, and to evaluate core-wide stability and regional stability which are stability different in oscillation modes (e.g., Patent Literature 2).
There is also known a technique to evaluate nuclear thermal hydraulic stability in consideration of parameters such as reactor core average neutron flux measurement values (APRM), delay corresponding to a heat-transfer time constant in fuel rods, and main steam flow rate measurement values (e.g., Patent Literature 3).
Further, there is known an apparatus adapted to perform prediction analysis of stability based on decay ratios and to issue an alarm when sequentially detected stability of the reactor core exceeds a predicted value (e.g., Patent Literature 4).
There is also known a technology to successively calculate deterioration indexes of nuclear thermal hydraulic stability in a boiling water reactor based on plant information, such as reactor core power distributions, reactor core flows, reactor core pressure and feed water temperature, and to issue an alarm when these indexes exceed preset values (e.g., Patent Literature 5).
In addition, there is known a method for monitoring stability based on neutron flux space mode distributions obtained by calculation of reactor core characteristic values using a large number of LPRM signals (e.g., Patent Literature 6).